ML23269A045
| ML23269A045 | |
| Person / Time | |
|---|---|
| Issue date: | 07/20/2023 |
| From: | NRC/OCIO |
| To: | - No Known Affiliation |
| Shared Package | |
| ML23269A046 | List: |
| References | |
| NRC-2021-000179 | |
| Download: ML23269A045 (1) | |
Text
MEMORANDUM FOR:
FROM:
SUBJECT:
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D, C. 20555 September 15, 1992 The Chairman Commissioner Rogers Commissioner Curtiss/
Cammi ss ion er Remick Commissioner de Planque James M. Taylor Executive Director for Operations POPULATION DENSITY COMPARISON N w On June 24, 1992, the staff briefed the Commission on proposed changes to the reactor siting criteria (10 CFR Part 100).
In a Staff Requirements Memorandum (SRM) dated July 8, 1992, the Commission requested the staff to provide a comparison of the population density in the area adjacent to nuclear power plants at the start of operation with the current population density.
The comparison was to be done for a representative sample of the sites. This memo is in response to that SRM.
The staff has examined population densities within 10 and 30 miles of the reactor for a selected number of sites. In order to examine the validity of population projections in general, sites were chosen where licensees had provided initial population data and projections for the year 1990 in their Final Safety Analysis Reports (FSAR).
This limited the number of sites reviewed for two reasons.
First, information regarding population projections was not generally requested of applicants prior to about the early 1970's.
As a result, the accuracy of long-term population projections (beyond about 20 years), cannot be directly verified. Second, in the normal course of updating their FSAR's, a number of licensees replaced the population data that was included at the time of plant startup with present data.
Ten sites (including two units at one site) were selected which met the above criteria. These sites are considered reasonably representative since they include both high and low population density sites and are located in all regions of the nation.
Actual population data for the year 1990 within 10 and 30 miles of the selected sites was developed from 1990 U.S. Census data for the States and individual counties.
Table 1 lists the sites reviewed in alphabetical order and-also gives the year of plant startup. The table also lists, for the areas within 10 and 30 miles, the population density (in persons per square mile) at plant start-up, 1990 actual density, and 1990 projections made at the time of plant start-up.
Table 1 also shows the actual annual population growth rate and the ratio of the actual population density to that projected at the time of plant start-up.
Average growth rates and average ratios of actual to projected population density for the selected sites are also shown.
A review of the table shows that the sites examined comprise both high as well as low population density sites. Indian Point represents the highest population density nuclear power plant site in the U.S., whereas North Anna represents one of the lowest.
For comparison purposes, the average population density for the contiguous 48 states of the U.S. is about 83 persons per square mile.
Within a distance of 10 miles of the site, population growth rates have ranged widely from a decrease of about 1.7 percent per year (Beaver Valley 2) to an increase of 3.3 percent per year (San Onofre 2). The average growth rate within 10 miles for the selected sites was 0.4 percent per year. This is to be contrasted with the annual average population growth rate for the nation of about 0.9 percent per year.
Population growth within a 30 miles radius of the site has shown a smaller variation. Further, the annual average growth rate of 0.64 percent per year for the selected sites is also closer to the national average growth rate.
Comparisons have also been made of the actual 1990 population densities within 10 and 30 miles to those that were projected at the time of plant startup.
These are shown in the table as ratios of actual to projected population densities. Figures 1 through 6 are also presented which show the ratio of the actual to projected population densities within 10 and 30 miles vs. three factors: the starting year of the plant, the initial population density and the population growth rate.
Enclosures:
Table 1 Figures 1-6 cc: SECY OGC ctor rations
Table 1 Population Density Comparison Density 1990 1990 0-10 mi.
0-10 mi.
Density 1990 1990 0-30 mi.
0-30 mi.
at start Actual Projected growth Ratio at start Actual Projected growth Ratio Start per sq.mi.
density density
- rate, actual/
per sq.mi.
density density
- rate, actual/
No.
SITE Year 0-10 mi.
0-10 mi.
0-10 mi.
pct/yr projected 0-30 mi.
0-30 mi.
0-30 mi.
pct/yr projected l
Arkansas 2 1980 86 83 99
-0.15 0.83 28 35 32 0.99 1.11 2
Beaver Valley 1 1976 394 400 409 0.05 0.98 568 500 599
-0.39 0.83 3
Beaver Valley 2 1987 450 400 462
-1.69 0.87 550 500 567
-1.38 0.88 4
Brunswick 1975 40 57 56 0.98 1.01 49 67 64 0.87 1.05 5
Byron 1985 68 61 75
-1.03 0.81 184 212 204 1.23 1.04 6
Indian Point 3 1970 696 761 1300 0.20 0.59 1410 1595 1939 0.27 0.82 7
Limerick 1 1980 498 592 551 0.75 1.08 1215 1138 1172
-0.29 0.97 8
North Anna 1 1978 25 37 35 1.36 1.06 53 68 79 0.88 0.86 9
SanOnofre2 1983 188 320 264 3.30 1.21 323 443 446 1.95 0.99 10 Surry 1972 196 189 514
-0.09 0.37 217 283 281 0.64 1.01 11 SL Lucie t 1976 291 365 397 0.70 0.92 55 115 119 231 0.96 Average=
0.40 0.88 Average=
0.64 0.96
1.3-.-----*
1.2-* **********.. ************"***
1.1 -< *...
1 0.9-***...........
0.8-0.7 0.6~-
0.5 0.4 RATIO OF ACTUAL TO PROJECTED VS. START YEAR --WITHIN 10 MILES
- /.<
/
0.3 ~--.-----r- --...-----...-----,----*-,-
- --,- -~--- ----.---~--
1970 1972 1975 1976 1976 1978 1980 1980 1983 1985 1987 START YEAR FIGURE 1
RATIO OF ACTUAL TO PROJECTED VS. START YEAR --WITHIN 30 M ILES 1.15~ ----- ----*---- -----~
1.1 -+*****
1.05 1 -+*******************........ */ ******
0.95 0.9 0.85-
./
_,.,,,,.~.,.**r.'..
0.8-L-~-~--
....-----.----**-*7**-
-,------.--- r-1970 1972 1975 1976 1976 1978 1980 1980 1983 1985 1987 START YEAR FIGURE 2
RATIO OF ACTUAL TO PROJECTED VS. INITIAL DENSITY WITHIN 10 MILES 1.3~ ------------
- -----~
1.2-+*-***************.. *.............................,................
1. 1 -+* -*-*****...... *....
0.9 0.8 0.7 0.6,.
0.5 0.4 0.3 O 100
,,..~.. -
200 300 400 500 POPULATION DENSITY FIGURE 3
' ~
600 700
RATIO OF ACTUAL TO PROJECTED VS. INITIAL DENSITY WITHIN 30 MILES 1.15--...----
1 1 -l***-*I*******.. ********................... -.......... *************.. ****************...
1.05+****.. *****
0_95..............
0.9--....,.....
0.85+----- ---**.......................
/
/
,I',.,.,*
_/,*
,/
/
/
0.8-r--
=~-
-,------,----r---
--J 0
200 400 600 800 1 000 1200 1400 1600 POPULATION DENSITY FIGURE 4
RATIO OF ACTUAL TO PROJECTED VS. GROWTH RATE WITHIN 1.0 MILES 1.3..,--------
1.2-+****** *** ** ***************************** ***
1.1 -+ *******-*****
1 *+******* ******
_ __ /
--~-.;;.~*-*
0.9*+********-* ***-************
- I
..... *******l
~-......*--
0.8-+*--*-*************
0.7-+**************** ** ***. ** *******************
- * ******* **I
- I 0.6*+**************.. ** ****-*- *********************............................
0.5 0.4-+**** *-******
... * ********I Q.3-4---- ~----~
-*-*-***--*-----T---- **-
- *-*---i---***
-2
-1 0
1 2
3 4
GROWTH RATE, PERCENT/YEAR FIGURE 5
RATIO OF ACTUAL TO PROJECTED VS. GROWTH RATE WITHIN 30 MILES 1.15~
1.1 -1***********.
1.05-+.......
1 *+****"***
0.95-+.............
0.9 -,........... *..
0.85-Q.8....,__ _ __ r ***-*-* ---*-7 ---- *--------T-- **--- *----1 ***-*--- -~-
- --*,*---**-~----
-1. 5
-1
-0. 5 0
- 0. 5 1
1. 5 2
2.5 GROWTH RATE, PERCENT/YEAR FIGURE 6
~
MEMORANDUM TO:
FROM:
SUBJECT:
P-t, ;(flJ UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 25, 1994 The Chairman Commissioner Rogers Commissioner de Planque James M. Taylor, Executive Director for Operations ERRATA PAGES FOR PROPOSED REVISION TO 10 CFR PARTS 100 AND PART 50 (SECY-94-194)
Attached are errata pages providing clarification and corrections, in line-in line-out version, on proposed revisions to 10 CFR Part 100 and 10 CFR Part 50 (SECY-94-194).
Two errata pages correct the Federal Register Notice (FRN),
and an errata page corrects Enclosure 5 of the paper.
The additional parenthetical phrase on page 8 of ~he FRN provides a technical clarification. The statement on page 9 of the FRN incorrectly states that the thyroid weighting factor of 0.03 is based on risk of cancer incidence. The weighting factor, in fact, is based upon risk of latent cancer fatality and genetic disorders.
An errata page for Enclosure 5 to SECY-94-194 also deletes this incorrect statement.
These corrections do not affect the validity of the analysis contained in nor do they alter the staff conclusion that the current dose criteria are equivalent to a value of either 27 or 35 rem total effective dose equivalent (TEDE), depending upon whether risk of latent cancer fatality or latent cancer incidence, respectively, is used..
Attachments:
As stated cc:
L. Soffer, RES 415-6574
i'
_)
.f"ission product release within*containment associated with major core damage, maximum allowable containment leak rate, a postulated single failure of any of
-the fission product cleanup systems, such as the containment sprays, adverse site meteorological dispersion characteristics, an individual presumed to be located at the boundary of the exclusion area at the centerline of the plume for two hours without protective actions), believes that this criteri-0n has clearly resulted in an adequate level of protection.
As an illustration -0f the conservatism of this assessment,_ the maximum whole body dose received by an actual individual during the Three Mile Island accident in March 1979, which involved major core damage, was estimated to be about 0.1 rem.
In the proposed rule, the Commission is proposing two changes in this area.
First, the Commission is proposing that' the use of different doses for the whole body and thyroid gland be replaced by a single value of 25 rem, total effective dose equivalent (TEDE).
The total effective dose equivalent concept is consistent with Part 20 of the Commission's regulations, and is defined as the deep dose equivalent (for external exposures) plus the committed effective dose equivalent (for internal exposures). The deep dose equivalent is the same as the present whole body dose, while the committed effective dose equivalent is the sum of the products of doses to selected body organs times weighting factors for each organ that are representative of the radiation risk associated with that organ.
The proposed use of the total effective dose equivalent, or TEDE, is based upon two considerations. First, since it utilizes a risk consistent methodology to assess the radiological impact of all relevant nuclides upon all body organs, use of TEDE promotes a uniformity and consistency in assessing radiation risk that may not exist with the separate whole body and thyroid organ dose values in the present regulation. Second, use of TEDE lends itself readily to the application of updated accident source terms, which can vary not only*with plant design, but in which additional nuclides besides the noble gases and iodine are predicted to be released into containment.
The Commission has examined the current dose criteria of 25 rem whole body and 300 rem thyroid with the intent of selecting a TEDE numerical value equivalent to the risk implied by the current dose criteria. These risks consist of the risk of developing cancer some time after the exposure (latent cancer. incidence), as well as a delayed risk of cancer fatality (latent cancer fatality).
For a dose of 25 rem whole body, the individual risk of latent cancer fatality is estimated to be about 2.5 X 10-2* the risk of latent cancer
- i:a=~:~j:f:'q:Fi::Yt'~t::i&:~::i:~::=:<:ifil::l::%:irtti~~iw!llill~~1:11,siitifi!~ti:~d11:;:i~~~i@:~:1:1,~l¥IIi!I~ th
'f'*f~k=*=*='~'f 1
Y~'!'nt'*=*='~'a'~~~'rAf,at'aYT'ty~*=*==y5===*=*=aho'~'t 2 ~r i~-3 ;
0
~~e 0 risk o~e~ ate~~
0
~a~cer e incidence is about a factor of ten higher.
If the risk of latent cancer fatality is selected as the appropriate risk measure to be used, the current do~e criteria represent a risk of about 2.7 X 10-2*
Using a risk coefficient of!about 10-3 per rem, the risk of latent cancer fatality implied by the ~urrent dose criteria is equivalent to 27 rem TEDE. (BEIR V estimates a latent cancer fatality risk coefficient of about 5 X 10-4 per rem,.if the dose is received over a period of days or more; however, if the exposure period is shorter, such as 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s~ the risk coefficient is approximately double.)
8
- I
,I ir
~
If latent cancer incidence rather than fatality were used, the current dose criteria would correspond to a value of about 35 rem TEDE.
The er~aA wei~htiA~ fa~ter fer the thyretd ~laAd iA Part 20 is 0.03, aAd is hased upeA ri~k ef.eaAeer iAeideAee.
UsiA~ this wei~htiA~ faeter, the eurreAt criteria ef 25 rem \\~hel e hedy aAd 300 rem thyreid res1:1lt iA a TEDE ef 34 rem (25 pl 1:1s 300 x 0.03).
The Commission is proposing to use the risk of latent cancer fatality as the appropriate risk measure since quantitative health objectives (QHOs} for it have been established in the Commission's Safety Goal policy. Although the current dose criteria are equivalent in risk to 27 rem TEDE, as noted above, the Commission is proposing to use 25 rem TEDE as thedose criterion for plant evaluation purposes, since this value is essentially the same level of risk as the current criteria.
Nevertheless, the Commission is specifically requesting comments on the use of TEDE.
Comments are requested on whether the current dose criteria should be modified to utilize the total effective dose equivalent, or TEDE, concept.
The Commission is also requesting comments on whether a TEDE value of 25 rem (consistent with latent cancer fatality}, or 34 rem (consistent with latent cancer incidence}, or some other value should be used.
- Finally, because the thyroid weighting factor is equal to a value of 0.03, there exists a theoretical possibility that an accidental release composed only of iodine could result in a TEDE less than 25 rem, yet result in a thyroid dose of over 800 rem.
Although the Commission believes that the likelihood that an actual accident would release only iodine is highly unlikely, comments are also requested as to whether the dose criterion should also include a "capping" limitation, that is, an additional requirement that the dose to any individual
- organ not be in excess of some fraction of the total.
The second change being proposed in this area is in regard to the time period that a hypothetical individual is assumed to be at the exclusion area boundary.
While the duration of the time period remains at a value of two hours, the Commission is proposing that this time period not be fixed in
- regard to the appearance of fission products within containment, but that various two-hour periods be examined with the objective that the dose to an individual not be in excess of 25 rem TEDE for any two-hour period after the appearance of fission products within containment. The Commission is proposing this change to reflect improved understanding of fission product release into the containment under severe accident conditions. For an assumed instantaneous release of fission products, as contemplated by the present rule, the two hour period that commences with the onset of the fission product release clearly results in the highest dose to a hypothetical individual offsite. Improved understanding of severe accidents shows that fission product releases to the containment do not occur instantaneously, and that the bulk of the releases may not take place for about an hour or more.
Hence, the two-hour period
. commencing with the onset of fission product release may not represent the highest dose that an individual could be exposed to over any two-hour period.
As a result, the Commission is proposing that various two-hour periods be examined to assure that the dose to a hypothetical individual at the exclusion area boundary wi 11 not be in excess of 25 rem TEDE over any two-hour peri.od after the onset of fission product release.
B. Site.Dispersion Factors Site dispersion factors have been utilized to provide an assessment of dose to an individual as a result of a postulated.
accident. Since the Commission intends to require that a verification be made 9
Risk Equivalence Between Current Dose Criteria and Total Effective Dose Equivalent CTEDE) Values
- 1.
It is desired to determine the numerical value of total effective dose equivalent CTEDE) corresponding to the current dose criteria of 25 rem whole body and 300 rem thyroid.
- 2.
To determine this, the risk imposed upon an individual as a result of receiving the current dose criteria (25 rem whole body and 300 rem thyroid) should be the same as for the dose in terms of total effective dose equivalent CTEDE).
- 3.
At this dose value, prompt fatality is precluded. There is a possibility of latent cancer incidence and as we 11 as latent cancer fatality.
However, the risk of latent cancer incidence is not the same as the risk of latent cancer fatality.
- 4.
The risks associated with the current criteria are as follows:
- For a dose of 25 rem whole body Risk of latent cancer fatality= 10*3 per rem* X 25 rem= 2.5 X 10*2 Risk* of latent cancer incidence= 5 X 10*2
- For a dose of 300 rem thyroid Risk of latent cancer fatality = 7 X 10*6 X 300 = 2.1 X 10*3 Risk of latent cancer incidence= 7 X 10*5 X 300 = 2.1 X 10*2
- 5.
Based on latent cancer fatality~ the current dose criteria represent a risk of 2.5 X 10*2 plus 2.1 X 10* = 2.7 X 10*2
- Using a risk coefficient of 10*3 per rem for both whole body exposure and total effective dose equivalent, the current dose criteria is equivalent to about 27 rem TEDE.
- 6.
Based on latent cancer incidence, the current dose criteria represent a risk of 5 X 10*2 plus 2.1 X 10*2 = 7.1 X 10*2
- Using a risk coefficient of 2 X 10*3 per rem for cancer incidence, the current dose criteria is equivalent to.about 35 rem TEDE.
The thyroid orgaR weightiRg factor of 0.03 giveR iR Part 20 is based upoR risk of caRcer iRcideRce aRd its use yields a value of 34 rem TE:DE: (25 pl us 0. 03 X 300), i R close agreemeRt with 35 rem TEDE.
From BEIR V !lni.:::::::~~1~:::::::~µµ1:1::P:i$:J.~:9P:i:I~, the risk is 5 X 10*
4 per rem if the dose is recei\\ied... o'\\ief""a"*pe*rlod... O'f.. *a* few days or more; if received over a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the risk is approximately twice that.
1
.P-;:: to'V
- UNITED STATesSECRETARIAT _RECORD 00'11 NUCLEAR REGULATORY COMMISSION wAsH1NGToN, D.C. 20555-0001 April 30, 1996 MEMORANDUM TO:
Chairman Jackson ~
FROM:
for Operations
SUBJECT:
PROPOSED RE IONS TO 10 CFR PART 100 AND PART 5 (SECY-94-194) - RESPONSE TO QUESTIONS In an April 9, 1996, memorandum, you raised several questions regarding the proposed revisions to 10 CFR Part 100 and Part 50 in order to clarify aspects of the proposed rule changes.
Where Research and NRR. disagree (i.e.,
timeframe for calculating dose), you requested the pros and cons to both viewpoints.
Attached are the staff responses to your questions.
As discussed during the April l, 1996, briefing, there is no disagreement regarding the majority of the issues involved in this rulemaking.
It is the issues regarding the timeframe for c~l~ulating the two hour dose (Question 2) where there is disagreement betweeh Research and NRR.
In response to this question, the pros and cons of both viewpoints are presented.
Attachment:
As stated cc:
Commissioner Rogers Commissioner Dicus SECYV".
Chairman's Questions and Responses 10 CFR Part 100 and Part 50 Revision Ql.
Will operating plants apply to use aspects of these new rules? Can the proposed regulations be written to allow use of. for example, the new source term - without the necessity of exemptions?
Al.
The proposed rules were intended to be forward-looking; they were developed for new applications for power reactor designs and sites.
As written, they are not applicable to operating plants.
However, it is likely that the industry will want to apply certain aspects of the rules to operating plants. Therefore, to the extent possible, the staff considered the impact of operating plants seeking to use aspects of the proposed rules.
The proposed rules could be re-written to be applicable to operating plants on a voluntary basis and, thus, precluding the need for exempti-0ns.
However, a revision would likely require publishing the
- rule for an additional period of public comments, thereby delaying issuance of the final rules.
Licensees already submitted applicati'ons to use the updated *source.:term insights, which is an important aspect of the revised rules.
NEI submitted ~ ~eneric framework document to the NRC in November 1995 that provides industry's proposal for applying the updated source term insights. The staff is reviewing this document and will present the results of the review in a Commission Paper.
Q2.
Reqardi nq the Difference of Opinion of "any" vs. first two hour timeframe for dose calculations.
Q2a.
What is the impact on risk by applying the new source term {S.T.)
during its worst two (2) hour period in the containment vs. the first two (2) hour period following onset of the source term release?
A2a.
There is essentially no impact on risk by applying the new source term during the worst two hour period vs. the first two hour
.period following the onset of the source term release.
For *plants with active spray systems, the difference between the first two hours versus the worst two hours is negligible.
The insights gained from risk studies performed by both industry and NRC.,
including NUREG-1150, demonstrate that the risk significant sequences are those where the containment fails or is bypassed early and not those where the containment is intact and leaking at its design basis leak rate, which is the assumption in the Part 100 dose evaluation.
The dose evaluations required under Part 100 (two hour dose at the exclusion area boundary and 30 day dose at the 19w population zone boundary) are a key part of the NRC's licensing review of nuclear power plants and provide acceptance criteria for judging, not only site suitability, but the effectiveness of plant design features that limit the release of fission products for design basis
2 accidents (OBA).
For advanced LWR designs that do not employ active fission product removal systems, such as containment sprays, the calculated OBA dose for the worst two hours could be higher than the first two hours by about a factor of two. (There is essentially no difference in the calculated dose between the worst two hours and the first two hours for designs with acti.ve fission product removal systems.).
As discussed above, the use of the worst two hours provides a more stringent test of a design's ability to limit the fission product release from OBAs.
However, the use of the worst two hours could lead to modifications in design (e.g., require a smaller containment leak rate, spray system or larger exclusion area boundary).
The risk reduction potential or defense-in-depth value of such changes is discussed below under the response to ~uestion 2b.
The issue for Commission consideration is.the extent that risk insights should be considered in selecting the dose evaluation period in accordance with the recent Commission Policy Statement on the *use of PRA Methods in Nuclear Regulatory Activities, specifically with respect to the following statements in the Policy Statement:
11 (1) The use of PRA *technology should *'be *increased in *all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional Hefense-in-depth philosophy.
(2) PRA and associated analyses should be used in regulatory matters, where practical within the bounds of t~e state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices."
- The staff believes that the source term postulated for use in the licensing evaluation should be one that provides a substantial challenge for purposes of judging the effectiveness of the _plant design to limit the fission product release from OBA.
Q2b.
Similarly, what is the impact on defense-in-depth, i.e.,
espe~ially containment design (~ressure and volume)?
A2b.
The design of the containment (pressure and volume) is governed by the temperature and pressure conditions associated with design basis accidents (double-ended LOCA or main steamline break}, and not by the dose criteria of Part 100.
As a result, the potential effect of either approach on the containment design pressure and volume would not be significant.
The dose criteria of Part 100 is focused on limiting the fission product release to the public from DBAs.
As discussed above,
3 there is essentially no difference between the calculated dose from either approach for plants with active fission product removal systems, and therefore, no impact on defense-in-depth.
For plants without active fission product removal systems, the impact of using the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> vs. the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> could be one or a combination of the following:
- 1) A smaller allowable containment leak rate,
- 2) A larger distance to the exclusion area boundary (EAB),
or
- 3) Addition of an active fission product removal system.
such as a spray or filtration system.
The impact on defense-in-depth (and risk) from either of the first two alternatives would not be significant.
The addition of an active fission product removal system, such as
- a spray~ could~ for some accident scenarios, provide some defense-in-depth, but would not impact those scenarios where the active fission product ~emoval system is not effective e.g., containment bypass and station blackout.
However, neither appro~ch will necessarily always lead to the addition of an active fission product removal system.
If the addition of such a system is deemed important for defense-in-depth, accident management or to compensate for ~ncertainties, staff believes that it should be addressed explicitly on its own merits for the design in question and not be implemented indirectly through use of the Part 100 dbse evaluation (see response to Question 4a).
Q2c.
Explain how the new S.T. can be "abused" and how is the rule structured to prevent this.
A2c.
While the new source term (NUREG-1465) provides specific updated source term characteristics for a PWR and a £WR, it also suggests that 11.[A]n applicant may propose changes in source term parameteis (timing, release magnitude, and chemical form) from those cohtained in this report, based upon and justified by design specific features."
The issue of 11 abuse 11 of the new source term, primarily arises from concerns that an applicant will attempt to adjust the release timing of the source term and, to a lesser extent, the release magnitude and plant response (fission _product removal rates) to ensure the calculated dose meets the dose criteria.
By attempting to extend the duration for the gap release phase and, thus, delaying the onset of the early in-vessel release, there is the concern that a significant fraction of the release that contributes to the EAB dos~ could be moved outside of a first two hour dose evaluation period.
The use of the worst two hour
4 dose evaluation period reduces the benefit of such "abuse" by identifying the dose for that two hour period that would result in the highest dose.
On the other hand, with a worst two hour evaluation period, an applicant could also propose to adjust the magnitude, timing, and removal rates to stay below the dose criteria.
Neither approach prevents an applicant from proposing such adjustments and likewise, does not eliminate the need for the staff to'understand the proposed design and its characteristics.
However, the use of the first two hours will place a greater burden on the staff to ensure that a licensee does not "abuse" the new source term.
In addressing how the rule is structured to prevent "abuse", it is important to remember that the source term used for the dose evaluation is a source term postulated for purposes of site analysis and assumes a substantial meltdown of the core.
The TID-14844 source term, which assumed an instantaneous release of fission products, was used for over thirty years.
Over ten years of research went into developing the NUREG-1465 source term.
It.
is expected that a considerable technical basis would be required to justify a source term significantly different than NUREG-146'5.
However, as discussed in response to Question 2a, with either approach, the staff believes that the source ter~ postulated for use in the licensing evaluation should be one that provides a substantial challenge for purposes of judging the effectiveness of the-~lant design to limit the fission product release from OBA.
The staff believes that the rule as proposed, is sufficient. to minimize the potential for "abuse" of the new source term.
- Q2d.
In the recent revisi6n to Appendix J, the staff recommended changes were based in part, on the risk significance of changes in containment leakage, i.e.. explain this viz-a-viz the two different staff positions.
A2d.
The risk from a plant is predominately from severe accidents -
large release due to early containment failure or bypass.
This risk is largely independent of containment leak rate until the leak rate starts *to approach the magnitude of major containment failure (many times higher that the current design containment le~k rate~). These risk insights were part of the basis for the changes in Appendix J, which increased the surveillance intervals for various types of containment leak rate testing, but not the allowable leak rate itself.
Q2e.
What is the most important insight regarding the new source term gained from all the extensive research performed on this subject?
What are the implications for this insight from applying the S.T.
during its worst vs. first two (2) hour period in the containment?
5 A2e.
The most important research insight regarding tne new source term is with regard to timing and that fissjon products are released into containment over a significant period of time involving several hours, rather than instantaneously. The time is associated with the phases of reactor core degradation during a severe accident and consists of several release phases, including the 11gap 11 and the "early in-vessel" phases.
In the 11gap 11 phase, the fuel cladding has failed and volatile fission products residing in the space or 11gap 11 between the fuel pellet and fuel cladding are released into containment. This period has a duration of o~s hours based upon review of representative severe accident sequences for, current designs.
In the early in-vessel" phase, fuel melting and relocation to the lower portion of the reactor pressure vessel takes place, but before any failure of the reactor pressure vessel. This results in a large release of volatile fission ~roducts as well as release of lesser quantities of non~volatile fission products iTito containment. This period has a duration of 1.3 and 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, for PWRs -and BWRs, respectively.
Although the new source *term remains consistent with the objective, noted in Footnote 1 to 10 CFR Part 100 that the source term be associated with a "substantial core meltdown with subsequent release of appreciable fission products", there are several implications from applying the source term during-its worst vs.
first two (2) hours.
First, ipplication of either the worst two.
hours or the first two hours will provide credit, due to radioactive decay, for designs that delay the initiation of fission product release into the containment.
However; the use of the first two hours will provide greater incentive by providing greater credit for designs that can provide additional delays in the in-vessel release phase.
If the period of gap release is extended to much beyond the first one half hour, then the staff is concerned that one of the basic objectives of having a significant rel~ase into the ~ontainment may not be satisfied.
In either case, the applicant must provide sufficient technical information to justify longer delays in the initiation or duration of the gap or in-vess~l release phases. Although the use of the first two hours provides greater incentive to designers to delay the release, it also results in a greater *burden on the staff to ensure that the applicant does not 11 abuse 11 the new source term insights.
In either case, the rule is structured to ensure that there would be a significant fission product release into the containment during the two hour period to provide for evaluating the effectiveness of the fission product removal capabilities of the plant.
Q2f.
If we select the 11 any 11 two hour option. should we re-address System 80+ to be consistent with this criteria?
6 A2f.
It would not be necessary to readdress System 80+ if we select the worst two hour approach.
System 80+ used the draft NUREG-1465 release characteristics and included a safety grade containment spray system.
As discussed above, with the containment spray system, the dose estimate under the first two hour approach is very nearly equal to the worst two hour value.
NRR performed OBA calculations for System 80+ using the framework of the proposed rule with the worst two hour approach; these calculations show that the System 80+ design, under its design certification licensing basis, bounds the proposed criteria.
Q3.
Background:
Th'e public (e.g.. TIME magazine, specific individuals who have raised safety questions) have been told that Part 100 is legally only for "siting". Yet, we have also stated that the NRC staff uses the dose guidelines of Part 100 as acceptance criteria in evaluating issues that may affect design basis accident dose consequences at operating reactors.
Is Part 100 truly only for siting? Should the rule be clarified to state that it -is also an ultimate limit for.operating reactors and postulated events (which appears to be.how it is applied)?
A3.
Although Par,t 100 is a reactor siting regulation, both the in-containment accident source term postulated to evaluate a site under Part 100, as well as the dose criteria of Part 100, are used by the staff {as.described in NUREG-0800, the Standard Review Plan) to evaluate the acce~tability of the mitigative capability of a plant design.
As part of reactor licensing, the staff evaluates the ability of a plant to cope with a number of postulated accident~, collectively referred to as design basis accidents.
Some of these are a postulated spent fuel handling accident, a postulated steam generator tube rupture, a postulated steam line break, and a postulated loss-of-coolant-accident (LOCA}.
The LOCA is usually the limiting design basis accident; the LOCA dose is.evaluated using the source term postulated for Part 100 and the do~e criteria pre*sented in Part 100.
Some applications flo~ing from it include setting the allow~bl~
containment leak rate, setting containment valve closure times, determining the allowable performance for fission product *C leanu*p systems, such as sprays and filters, determining the "drawdown" time in which a secondary containment annulus must reach a slight negative pressure, setting the radiological accident environment used to evaluate control room habitability, aDd determining the radiological environment for equipment qualification.
In addition to using the dose criteria to judge the original plant design at licensing, the dose criteria are used to assess whether the plant continues to meet its design basis, and to evaluate changes to the design proposed by the licensee during the life of the plant. This is why the proposed rule changes relocates the dose criteria to Parts 50.
r
7 Q4.
Policy Issues Q4a.
Should the policy issue of the ~roper mix of preventative and.
- mitigative design features be brought forward with this rule?
A4a.
The staff does not believe that the policy issue of the proper mix of preventative and mitigative design features needs to be brought forward with this rule.
The Commission's August 8, 1985, Policy Statement on Severe Reactor Accidents Regardin~ Future Designs and Existing Plants articulates the issue of balancing accident prevention and mitigation features, neither to the exclusion of the other.
In addition, in SECY-88-203, "Key Licensing Issues Associated with DOE-Sponsored Advanced Reactor Designs, 11 dated July 15, 1988, the staff stated that the criteria it proposed to use to implement the Advanced Reactor Policy Statement would allow a trade off between plant protection and accident mitigation to achieve an equivalent level of safety.as current generation LWRs.
However, in implementing this policy, the staff intends to bring forward in a separate paper, the issue regarding the need for a containment spray system as well as other policy issues for the AP 600 design.
In addition, in response to the SRM dated.September 14, 1993, staff will be providing a recommendation on the need for generic rulemaking to addresi severe accident performance for advanced reactors. This recommendation is currently scheduled to be provided to the Commission in December 1996.
As part of any generic rulemaking effort to address severe accident performance, alternatives to the dose criterion could also be studied that would ensure both early public protection for individuals close to the plant and provide a less prescriptive way of ensuring effective fission product mitigation by specifying a release limit to the environment.
Q4b.
Are there any other policy issues that are related to these rule changes; and. have any issues changed significantly since the last
- public comment period?
A4b.
Although no other issues have changed significantly since the last public comment period, there is -0ne additional policy issue that relates both to this rule change and to the use of the worst two hours vs. the first two hours.
This additional policy issue, which was raised as a result of public comments, and is related to this rule change is "what is the purpose of the two hour EAB dose evaluation period".
The current Part 100 requires the evaluatiori of the EAB dose for a two hour period following the onset of the postulated release.
With the use of the TID-14844 source term the worst two hour period is the same as the first two hour period.
Staff has not been able to find a clear rationale for the purpose of the two
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period as applied in the original Part 100.
With the introduction of the NUREG-1465 source term, the question has arisen as to "what should be the purpose of the two hour EAB dose evaluation period".
The issue for Commission consideration is whether the two hour period should provide for limiting the early releases from the plant (first two hours) or should it be for the purpose of limiting the maximum release rate (worst two hours).
In either case, the staff will need to ensure that the postulated fission product release represents a significant challenge for evaluating the plant design.
- Q5.
Seismic Issues QSa.
Do our Regulatory Guides prescribe a minimum acceptable EAB distance and maximum population density? If so, do we envision siting a plant if these "guidelines" couldn't be met?
ASa.
Our Regulatory Guides do not prescribe a minimum acceptable EAB distance, nor a maximum population density.
Guidance on the exclusion area and population density are.given in Revision 1 of Regulatory Guide 4.7,.. "General *site Suitability Criteria for Nuclear Power Stations", issued in 1975. This guide does not prescribe a minimum acceptable distance for the EAB, but states that a distance of 0.4 miles, in combination with plant design features, will usually be found to be able to meet the dose criteria of Part 100. Similarly, the Guide does not contain a maximum acceptable population density, but states that if the population density exceeds 500 people per square mile out to a distance of 30 miles, or is projected to exceed a population density* of 1000 persons per square mile at the end of pl ant life, the applicant should present a special analysis of alternative
.sites having lower population densities.
The staff proposes to revise Regulatory Guide 4.7 in conjunction with the revised rule..
The revised Guide would. contatn no discussion of minimum exclusion area distance, other than it should be of such a size that it satisfies the dose criteria.
Since the Guide also indi~ates that a reactor site should be amenable to the development of an adequate security plan, and notes that a distance of about 100 meters will normally be consi~ered adequate for such a purpose, there i~ an implication that the distance to the exclusion area boundary should be at least 100 meters. This distance, it should be pointed out, is significantly smaller than the smallest EAB distance, which is 277 meters, for any operating plant. Most EAB distances are 400 meters or larger.
Similarly, the proposed revision of Regulatory Guide 4.7 states that sites having population densities of 500 people per square mile out to a di stance of 20 mil es are "preferred". The Guide indicates that sites having a higher de~sity, so long as they are
9 "away from" very densely populated centers could be found acceptable depending upon a showing that the higher population density site had advantages in safety, environmental or economic consideratio.ns. Therefore, the regulatory guide provides the Commission flexibility to consider a range of parameters in its siting decisions, and not be bound to apply numerical guidance in a rigid fashion.
Q5b.
If it were not for QBE being mentioned elsewhere in the regulations, would the staff recommend "evolving" to the SSE exclusively?
A5b.
The staff does not currently recommend "evolving" to the exclusive use ~f the SSE and does not anticipate making such a recommendation in the future.
At the level of the SSE ground motion plants should be able to shut down and remain shut down until aut~orized to start up, but at the ground motions due to QBE plants *should be able to continue to operate.
There are two aspects to ~arthquake resistant designs: one aspect ensures that a rare earthquake (SSE) will not create failures by demonstrating that allowable stress and deformation 1 imits are met, the other aspect deals with the assurance that smaller but more frequent earthquakes (QBE) do not push a component beyond its endurance or fatigue limit by demonstrating adequacy of design through fatigue evaluation.and shake table tests using applicable dynamic and seismic loads.
~hile the primary aspect of public safety considerations can be treated with SSE design criteria alone, considerations to assure safe continued operation (the QBE) must also enter into plant design.
Consequently, the QBE is likely to remain in the Codes and Standards used for plant design to withstand th~ effects of earthquakes that might prudently be expected during the plant*lifetime.
It can be argued that a fraction of the SSE can be used as a surrogate for the QBE, but that will obscure the true purpose of the design process of ensuring safety against vibratory mo ti on from more frequent earthquakes.
Experience has shown that the earthquake resistant design process discussed above has provided satisfactory perfnrmance in actual earthquakes.
Though the choice of an QBE 1eve1 wi 11 be an economic one made by the owner, the requirement for shutdown and thorough inspection to assure that there has not been any permanent damage or deformation of critical structures, systems, or components, particularly the active parts in the plant, if the QBE has been exceeded, is a safety concern of interest to the staff.